External Radiation Dose Calculator - HELP(Virtual Geiger Counter)

The External Radiation Dose Calculator determines the radiation dose from a shielded gamma source. The source can be a point source, or a cylindrical volume source with an evenly distributed concentration of radionuclides. The shield may consist of consecutive layers, each of which may also contain additional radionuclides.
For source and each shield layer, a number of common materials and compositions of natural radionuclides can be selected, or a custom mix of elements and radionuclides can be entered.
The receptor location can be varied in two dimensions.

Typical situations covered by the calculator:

The Calculator does take into account:

Nuclide-specific Gamma radiation for a selection of important natural and a few artificial radionuclides.

in case of a volume source, the Gamma radiation emitted from the top of the cylinder.

Dose contributions from the source, plus from radionuclides contained in the shield material(s), if applicable

Self-shielding within a volume source, and within shield layers containing radioactive materials.Note: Self-shielding can also be taken into account for sources otherwise treated as point sources, if sufficient material properties are known (density and attenuation coefficients); in this case, the effective radiation emitted from the top of a cylinder with height equaling diameter is used as the activity of the point source, and the source is labeled "Effective Point Source".

Attenuation of Gamma radiation through a selection of common shield materials, also when arranged in multiple layers. The calculator uses the point-kernel method with buildup factors (see Calculation Details).

Gamma attenuation in air

Cosmic ionizing radiation, if required (calculated according to UNSCEAR 2000)

For a residential exposure scenario, the radiation dose and risk for an individual living on soil contaminated from uranium and/or in a home built
from such contaminated material can be determined with the Uranium in Soil and Building Material Individual Dose Calculator. It not only covers external exposure, but also the following pathways: ingestion of soil, inhalation of fugitive dusts, ingestion of produce grown in the soil, ingestion of contaminated
groundwater, and inhalation of radon.

The geometry of the situation and the properties of the materials involved are defined in the Input Data table.

With a HTML 5-enabled browser, a section of the geometry of the situation is shown in a Graph.

Upon execution of the calculation, the graph shows the Gamma dose rate at the selected receptor location. The receptor can easily be moved to other locations by mouse clicking.

In addition, the gamma dose can be visualized in a color map, if the Show color map box is checked.Note 1: Each square in the color dose map represents the dose value in its center, not the value averaged over the square. Due to the sharp dose increase near point sources, the color scheme may change considerably with a decrease of the raster width, therefore.Note 2: In logarithmic mode, the color of the square with the highest dose value is red. The span given by the selected number of decades is represented by the rainbow colors red - yellow - green - cyan - blue - violet. If the range of values is not covered by the number of decades selected, the remaining (too low) values are displayed in a linear fade-out of violet.

Computing time increases considerably with number of shields, number of elements and radionuclides per layer, and in particular with color map enabled. The example parameter sets take a few seconds to compute on current machines.

The Result field repeats some important input data, reports any warnings about missing input data, and shows the calculation results for the current receptor location. The contributions from the source and from each layer containing radionuclides are shown separately and in summary.
The contents of the Result field can be highlighted and copied for further use.Note: The figures are for the current geometry - so, if the effect of a shield is to be compared to the open source, disable the shield (select Layer OFF) and compare the results manually.

The contents of the database of the calculator (decay energies of radionuclides, gamma attenuation and energy absorption coefficients) can be viewed with the Query Database button.

Notes:
This selection must be made before any other entry, since it resets the complete calculator!
After a selection is made, the calculator can be bookmarked to obtain future direct access to the mode selected.

Number of Shield layers

Point / Volume Source

Example data sets

These data sets preset the whole calculator for certain typical cases of interest. A first calculation result can then be obtained by just clicking the Calculate button. After selection, the example data can also be modified as required, to study the effect of parameter variations.Note: set Number of Shield layers=1 and select appropriate source mode first, before choosing an example data set!

Ex. 1: Shielding in free air(for Point Source Mode)

Study the impacts of distance, shield thickness, and shield material. With the highest Result Detail setting, the factors affecting attenuation can be checked for each single energy emitted by the radiation source.Note: when entering individual source radionuclides, make sure that the calculator's database actually contains the gamma energy data for these nuclides (Query Database button): for space considerations, the calculator contains energy data for a few selected nuclides only.

Ex. 2: 48Y Cylinder with Heels from UF6_nat(for Point Source Mode)

After unloading of an uranium hexafluoride (UF6) cylinder by heating in an autoclave, the decay products of the uranium remain in the cylinder as so-called Heels. See, how an "empty" cylinder emits much more gamma radiation than a full one.Note 1: This data set conservatively assumes that no residual UF6 remains in the cylinder.Note 2: In addition, the cylinder emits Bremsstrahlung which is not covered by this calculator.

Ex. 3: 30B Cylinder with Heels from UF6_enr(for Point Source Mode)

After unloading of an uranium hexafluoride (UF6) cylinder by heating in an autoclave, the decay products of the uranium remain in the cylinder as so-called Heels. See, how an "empty" cylinder emits much more gamma radiation than a full one.Note 1: This data set conservatively assumes that no residual UF6 remains in the cylinder.Note 2: In addition, the cylinder emits Bremsstrahlung which is not covered by this calculator.

Ex. 4: DU bullet buried in soil (for Point Source Mode)

Depleted uranium bullets buried in soil are difficult to locate by their gamma radiation emission.

Ex. 5: Uranium mill tailings cover(for Volume Source Mode)

Gamma radiation from uranium mill tailings can be reduced by soil covers, often applied in several layers. Check the effects of layer material, cover thickness and of residual radioactivity contained in the cover material.

Ex. 6: 48Y Cylinder with UF6_nat(for Volume Source Mode)

Determine the gamma dose rate outside the top of a typical transport cylinder for natural uranium hexafluoride.Note 1: This data set assumes that the cylinder is completely filled with solid UF6, which it normally isn't. Modify the Source density rhoso to 2.897 g/cm3 to account for the regular partial filling state.Note 2: In addition, the cylinder emits Bremsstrahlung and neutron radiation which are not covered by this calculator.

Ex. 7: 30B Cylinder with UF6_enr(for Volume Source Mode)

Determine the gamma dose rate outside the top of a typical transport cylinder for enriched uranium hexafluoride.Note 1: This data set assumes that the cylinder is completely filled with solid UF6, which it normally isn't. Modify the Source density rhoso to 2.76 g/cm3 to account for the regular partial filling state.Note 2: In addition, the cylinder emits Bremsstrahlung and neutron radiation which are not covered by this calculator.

This drop down list allows to easily disable parts or all of the properties of the source layer for test purposes:

NORMAL

the layer is fully operational

RAD. ONLY

the attenuation properties of the layer are disabled

ATTEN. ONLY

any radiation emission from the layer is disabled

Layer EMPTY

attenuation and radiation properties are disabled; the layer is filled with air

Layer OFF

the layer is completely removed

Total amount / Individual amounts below(Point Source only)

Choose, whether the composition of the source material is to be entered as total mass plus activity or mass concentrations, or as individual activities and masses.

Total amount: Mass figure(Point Source only)

Enter number

Total amount: Mass Unit(Point Source only)

Select unit from pick list

Source Material

Material data for the radiation source
Select sample material from the pick list (and modify any entries, if required), or enter new data in the table.
The pick list contains elemental compositions, as well as radionuclide compositions and radionuclide series.Note: The decay energy and attenuation data can be viewed with the "Query database" button.

Check box and enter density value to take self-shielding within the point source into account: if attenuation coefficients are available for the source material, then the activity of the point source is taken as the activity at the top surface of an upright cylinder with diameter equalling height, and the source is labeled "Effective Point Source".

rhoso - Source density [g/cm3](Volume Source only)

Value must be larger than zero.

Element / Nuclide

Enter short names of elements (e.g. Th) or radionuclides (e.g. U-238) and associated concentrations in weight-percent, or, for radionuclides, alternatively the activity concentration in Bq per gram of source material. For point sources, there is also the possibility to enter individual masses in grams and activities in Bq directly, if the box "Individual amounts below" is checked.
In addition, the short names of a number of pre-defined radionuclide compositions (e.g. U_nat) and radionuclide series (e.g. U-238++) can be entered.Note 1: For elements, the natural isotopic abundance of the element is assumed, if there is one defined (check with the "Query database" button) - otherwise enter individual nuclides.Note 2: For radionuclides, check the availability of the associated decay data with the "Query database" button.Note 3: Mass and activity concentrations entered for a radionuclide composition refer to the total mass/activity of all nuclides of the nominal element contained (e.g. U-238, U-234, and U-235 for U_nat++), while those entered for a radionuclide series refer to the mass/activity of the nominal nuclide only (e.g. U-238 for U-238++).

Data import: Longer lists of input data can be imported by pasting the data to the "Results" field first, then clicking the "Import" button. For this purpose, the data must be delimited by space, comma, tab, or new lines, and "wt_%" must be encoded as 0, "Bq/g" as 1. So, direct import from spreadsheet applications such as Excel is possible by copying and pasting, if the data is organized in three columns for name, value, and unit.Note: make sure that you get decimal points (not decimal commas!) from your spreadsheet software!

Layer usage selector

This drop down list allows to easily disable parts or all of the properties of each shield layer for test purposes:

NORMAL

the layer is fully operational

RAD. ONLY

the attenuation properties of the layer are disabled

ATTEN. ONLY

any radiation emission from the layer is disabled

Layer EMPTY

attenuation and radiation properties are disabled; the layer is filled with air

Layer OFF

the layer is completely removed

Shield Material

Material data for each shield
Select sample material from the pick list (and modify any entries, if required), or enter new data in the table.
The pick list contains some elements, as well as elemental compositions and radionuclide compositions.Note: The attenuation data and/or decay energy data can be viewed with the "Query database" button.

rhoshn - Shield density [g/cm3]

Value must be larger than zero, for the shield to be effective.Note: If no value, or zero, is entered, this layer is treated as vacuum.

Element / Nuclide [wt_% / Bq/g]

Enter short names of elements (e.g. Pb) or radionuclides (e.g. U-238) and associated abundance in weight-percent, or, for radionuclides, alternatively the activity concentration in Bq per gram of source material. In addition, the short names of a number of pre-defined radionuclide compositions (e.g. U_dep) and radionuclide series (e.g. U-238++) can be entered.Note 1: For elements, the natural isotopic abundance of the element is assumed, if there is one defined (check with the "Query database" button) - otherwise enter individual nuclides.Note 2: For radionuclides, check the availability of the associated decay data with the "Query database" button.Note 3: Mass and activity concentrations entered for a radionuclide composition refer to the total mass/activity of all nuclides of the nominal element contained (e.g. U-238, U-234, and U-235 for U_nat++), while those entered for a radionuclide series refer to the mass/activity of the nominal nuclide only (e.g. U-238 for U-238++).

Data import: Longer lists of input data can be imported by pasting the data to the "Results" field first, then clicking the "Import" button. For this purpose, the data must be delimited by space, comma, tab, or new lines, and "wt_%" must be encoded as 0, "Bq/g" as 1. So, direct import from spreadsheet applications such as Excel is possible by copying and pasting, if the data is organized in three columns for name, value, and unit.Note: make sure that you get decimal points (not commas!) from your spreadsheet software!

The geometry parameters can be initialized with 5 predefined data set examples, corresponding to the example buttons in the Mode section.Note: While the buttons in the Mode section initialize all parameters, here only the geometry is affected.

Enter either y or b. The value must be larger than zero.
Upon entry of y, b is erased, and vice versa.Note: The x-axis runs through the point source, the top of the effective point source, or the top of the volume source, respectively.

Enter either r or sa. The value must be larger than zero.
Upon entry of r, sa is calculated automatically, and vice versa.

rs - Shield radius [cm] - or -sas - Shield surface area [m2]

Enter either rs or sas. The value must be equal or larger than the source radius.
Upon entry of rs, sas is calculated automatically, and vice versa.(In the case of a point source, this parameter is only required if the shield contains radionuclides)

integration step width [cm]

Maximum horizontal size of the section of the volume elements used for the point-kernel method.Note: smaller sizes improve the precision, but increase computing times and memory requirements considerably(In the case of a point source, this parameter is only required if the shield contains radionuclides)

integration step height [cm]

Maximum vertical size of the section of the volume elements used for the point-kernel method.Note: smaller sizes improve the precision, but increase computing times and memory requirements considerably(In the case of a point source, this parameter is only required if the shield contains radionuclides)

Select from pick listNote: The primary unit calculated is Gy/h. All other units are derived from this one.

Dose Options: Exposure for annual dose rates

Select occupancy form pick list

Dose Options: Receptor Material

Select from pick listNote: The absorbed gamma energy dose usually is calculated for air, and any further dose figures are derived from this value.

Dose Options: Terrestrial gamma dose coeff. in air [Sv/Gy]

Conversion coefficient from absorbed dose in air to effective dose equivalent for terrestrial gamma rays.
UNSCEAR (2000) recommends 0.7 Sv/Gy for adults, 0.8 for children, and 0.9 for infants.Note 1: this coefficient is used energy-independentlyNote 2: this coefficient is only used for receptor air, otherwise unity is usedNote 3: the coefficient used for the contribution from cosmic ionizing radiation is unity

Dose Options: Use buildup factors for:

Check to enable computing of buildup factors to compensate for the non-ideal geometry of the situation.Note: although the use of buildup factors increases computing time, unchecking is not advisable, except for test purposes.
For each layer, the material to be used for the buildup factor calculations can be selected manually from the respective drop-down list, or "auto select" can be chosen.Note: For "auto select", the buildup data for the layer material is used, if available; otherwise the buildup data for an estimated effective atomic number of the layer material is used.

Graph Detail: Show integration grid

Check to show the grid defined by the integration step width/height entered under Geometry Parameters

Graph Detail: Show dose color map

Check to calculate dose values for the following raster of x and y positions and show the result as a coloured map.Note: If this checkbox is checked, computing time increases considerably, in particular for volume sources and/or shields containing radionuclides!

Graph Detail: Raster width [pixel]

Raster width in pixels for calculation of dose values in color map
Min. value: 10 pixel
If no value, or zero, is entered, the minimum raster width is used.Note: Reduction of the number increases computing time considerably!

Graph Detail: Logarithmic color scale

Check for logarithmic color scale, otherwise a linear scale is used

Graph Detail: Decades

Number of decades to be covered by color map in logarithmic mode.
If no value, or zero, is entered, an appropriate scale is selected automatically.Note: If, for logarithmic scale, the range of values is too large for the number of decades selected, the remaining (too low) values are displayed in a linear fade-out of violet.

Result Detail: Dose from each layer

Select desired level of detail for the dose summary in the Result field.(The selections "by Radiations" are available in Point Source mode only and affect the contributions from the point source itself only)

a) based on Soil_US, density and/or radionuclides modifiedb) based on Utail_nor, density and/or radionuclides modifiedb) based on Utail_dgo, density and/or radionuclides modifiedNote 1: Mass and activity concentrations entered for an element composition refer to the total mass/activity of all nuclides contained, except for those only contained in the decay series indicated by "+".

Note 1: Mass and activity concentrations entered for a radionuclide composition refer to the total activity of all nuclides of the nominal element contained (e.g. U-238, U-234, and U-235 for U_nat++).Note 2: Radionuclide compositions are calculated with the material properties of the nominal element.

Note 1: Mass and activity concentrations entered for a radionuclide series refer to the activity of the nominal nuclide only (e.g. U-238 for U-238++).Note 2: Radionuclide series are calculated with the material properties of the nominal nuclide.

With these Radionuclide Series, the decay series can be composed as follows, for example:

For point sources, the attenuation of a radiation beam is calculated, as it passes through the shield (which may be composed of consecutive shield layers). For each decay energy of each radionuclide in the source, the energy-specific attenuation on the beam's path through the shield is taken into account. For sections of the path outside of the shield, attenuation by air is assumed.
If no attenuation data is stored for a shield layer material, it is calculated from the layer's elemental composition, if given, and if attenuation data is stored for the relevant elements.
Shields containing radionuclides are treated as additional volume sources (see below).

To compensate for the non-ideal geometry of the situation, buildup-factors are used for the shield layers. In case the material of a shield layer consists of more than one element, and buildup factors for the material are not available, an energy-specific effective atomic number (Zeff) is calculated for the layer material according to [Singh 2003], which is then used to determine the buildup factors. In case the buildup factors for a layer element (real or effective) are not available, the buildup factors are interpolated between those known for the elements with the nearest atomic numbers. For multi-layer shields, buildup factors are calculated according to the empirical formula provided in [Broder 1963].

For volume sources, the cylindrical source is divided into small volume elements in the form of rings with a rectangular section. This allows to handle the three-dimensional problem in two dimensions to save computing time. For each volume element, the attenuation of a radiation beam is calculated, as it passes through the rest of the source and through the shield (which may be composed of consecutive shield layers). For each decay energy of each radionuclide in the source, the energy-specific attenuation on the beam's path is taken into account. For sections of the path above the shield, attenuation by air is assumed.
If no attenuation data is stored for a layer material, it is calculated from the layer's elemental composition, if given, and if attenuation data is stored for the relevant elements.
If the receptor is horizontally displaced away from the common axis of the source and shield cylinders, the volume rings follow the displacement, and only their intersections with the source/shield cylinders are taken into account.
Shields containing radionuclides are treated as additional volume sources.

To compensate for the non-ideal geometry of the situation, buildup-factors are used for each layer. In case the material of a source or shield layer consists of more than one element, and buildup factors for the material are not available, an energy-specific effective atomic number (Zeff) is calculated for the layer material according to [Singh 2003], which is then used to determine the buildup factors. In case the buildup factors for a layer element (real or effective) are not available, the buildup factors are interpolated between those known for the elements with the nearest atomic numbers. For multi-layer situations (i.e. source plus at least one shield), buildup factors are calculated according to the empirical formula provided in [Broder 1963].